JP5749597B2 - Fast breeder reactor core - Google Patents

Description

  The present invention relates to a fast breeder reactor core.

  The core and fuel assembly of the fast breeder reactor are described in Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics” (Tohoku University Press, October 30, 2003, p279-286) and JAEA-Evaluation 2009-003. 37 and 43, 2009. The fuel assembly loaded in the core of a fast breeder reactor (FBR) is a bundle of fuel rods in which a plurality of fuel rods containing a plurality of fuel pellets containing depleted uranium enriched in plutonium (Pu) are bundled. A trumpet tube having a regular hexagonal cross section surrounding the fuel rod bundle, a coolant outlet portion located above the fuel rod bundle, and a neutron shield and coolant inlet portion (entrance) located below the fuel rod bundle. Nozzle).

  The core of the FBR is loaded with the plurality of fuel assemblies described above. A standard homogeneous core of the FBR includes a core fuel region having an inner core fuel region and an outer core fuel region surrounding the inner core fuel region, and a radial blanket region surrounding the core fuel region, specifically the outer core fuel region. Have. The Pu enrichment in the outer core fuel assembly loaded in the outer core fuel region is higher than the Pu enrichment in the inner core fuel assembly loaded in the inner core fuel region. Thereby, the power distribution in the radial direction in the core fuel region is flattened.

  As a form of nuclear fuel material contained in each of the inner and outer fuel assemblies, metal fuel, nitride fuel, oxide fuel, and the like have been studied so far. Of these nuclear fuel materials, oxide fuels are the most abundant. A fuel rod used for each of the inner and outer fuel assemblies has a fuel region formed in the central portion in the axial direction, and axial blanket regions formed above and below the fuel region, respectively. The fuel region has a length in the fuel rod axial direction of about 80 to 100 cm and is filled with a plurality of fuel pellets made of mixed oxide fuel (MOX fuel) in which Pu and oxides of depleted uranium are mixed. . The axial blanket region is filled with a plurality of fuel pellets made of depleted uranium oxide fuel. The core fuel region including the inner core fuel region and the outer core fuel region is a region where a large number of fuel pellets made of the above-described MOX fuel are arranged in the axial direction and the radial direction of the core.

  The blanket fuel assembly loaded in the radial blanket region has a plurality of fuel rods filled only with a plurality of fuel pellets made of depleted uranium oxide fuel. In the radial blanket region and the axial blanket region, among the neutrons generated by the fission reaction of fissionable material (for example, Pu-239) contained in the MOX fuel in the core fuel region, neutrons leaking from the core fuel region are U- It is absorbed by 238, and Pu-239, which is a fissile nuclide, is generated. These blanket regions contribute to Pu growth (growth ratio> 1.0) throughout the core.

When the FBR is started and stopped, and when the reactor power is adjusted, the control rod is operated. The control rod is configured by housing a plurality of neutron absorption rods filled with a plurality of pellets of boron carbide (B ₄ C) in a trumpet tube having a regular hexagonal cross section. As the control rods, a plurality of main furnace stop system control rods and a plurality of after-furnace stop system control rods are provided independently. The FBR can be urgently stopped with only one of the main furnace stop system control rod and the after-furnace stop system control rod.

  The burnup of each blanket region is lower than that of the core fuel region. When the isotopic composition of Pu when the fuel assembly is taken out from the core is compared between the core fuel region and each blanket region, the ratio of fissile nuclides is higher in the latter region than in the former region.

  M. SAITO is a nuclear fuel used for fuel assemblies loaded in various nuclear reactor cores in "Advanced Core Concepts with Enhanced Proliferation Resistance by Transmutation of Minor Actinides", Proceedings of GLOBAL 2005, Paper No. 172, 2005. Minor actinide (hereinafter referred to as MA) is added to the material, and the ratio of Pu-238 in the total Pu contained in the nuclear fuel material extracted from the core is tentatively increased to 20% or more. Is proposing a reactor-grade approach. On the other hand, B. Pellaud, in “Proliferation aspects of plutonium recycling”, CR Physique 3, pp.1067-1079, (2002), is a nuclear grade nuclear fuel material in which the proportion of Pu-240 is 18% or more. It is defined as

  Generally, when MA is added to the nuclear fuel material used in the fuel assembly loaded in the core fuel region in the core of the FBR, the reactivity coefficient related to the safety of the core such as the void reactivity and the Doppler coefficient changes. P. Of "Review Separation Conversion Engineering" (Separation Conversion Engineering Special Committee Japan Atomic Energy Society, February 2004). Fig. 3.1.9 of 3-5 shows the relationship between the MA addition rate and the Na void reactivity when MA is added to the nuclear fuel material contained in the fuel assembly loaded in the core fuel region of the fast reactor. Has been. From this figure, it can be seen that the void reactivity increases as the MA addition rate increases.

On the other hand, in the transition period from the light water reactor to the fast reactor, FBR using Pu obtained by reprocessing the spent nuclear fuel material of the light water reactor is introduced. In Japan's standard scenario, the introduction of FBR started around 2050, and it is considered that all light water reactors will be replaced with FBR in about 60 years. Fig. Of "Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan" by K. FUJIMURA et al., Journal of Power and Energy Systems, Vol. 3, No. 1, pp. 136-145, (2009). Fig. 5 shows the fuel composition of transuranic elements (TRU), which is the sum of Pu and MA of spent nuclear fuel materials of various light water reactors. From the figure, it can be seen that in the case of UO ₂ fuel, the greater the burnup, the more the MOX fuel is mixed with MAO than the UO ₂ fuel, and the lower the proportion of fissile Pu. When TRU is used as nuclear fuel for fast reactors, if the weight ratio of fissile Pu in the whole Pu is small, the weight ratio of TRU to be added (= TRU enrichment) increases, so the mixing ratio of MA further increases. growing.

Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics”, Tohoku University Press, pp. 279-286, October 30, 2003 M. SAITO, “Advanced Core Concept with Enhanced Proliferation Resistance by Transmutation of Minor Actinides”, Proceedings of GLOBAL2005, Paper No. 172, 2005 B. Pellaud, “Proliferation Aspects of Plutonium Recycling”, C. R. Physique 3, pp.1067-1079 (2002) “Review Review Separation Conversion Engineering”, Special Committee on Separation Conversion Engineering Japan Atomic Energy Society, pp. 3-4 to 3-5, February 2004 K. FUJIMURA, et. Al., “Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan”, Journal of Power and Energy Systems, Vol.3, No.1, pp.136-145, (2009)

  Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics”, Tohoku University Press, pp. 279-286, in the standard FBR homogeneous core described on 30 October 2003, the fraction of fissile nuclides in each axial blanket region above and below each fuel assembly loaded in the core fuel region Pu with a large amount of is generated. For this reason, in the nuclear fuel material in the upper and lower axial blanket regions present in the spent nuclear fuel assemblies taken out from the core fuel region, the ratio of fissile nuclides, that is, odd odd nuclides of Pu increases. In other words, there are few even nuclides of Pu contained in the nuclear fuel material in the axial blanket regions above and below the spent nuclear fuel assembly. When this nuclear fuel material is dissolved in the dissolution tank of the nuclear fuel reprocessing facility, the amount of nuclear fuel material existing in the axial blanket region should be reduced when criticality is considered. Don't be. Therefore, the time required for reprocessing spent nuclear fuel material becomes longer.

  Reduce the percentage of Pu fissile nuclides (odd nuclides) contained in each spent nuclear fuel material present in all regions of the spent nuclear fuel assemblies extracted from the core fuel region of the FBR. If the level of Pu contained in the reactor can be a reactor grade with a high proportion of even-numbered nuclides of Pu, the time required for reprocessing spent nuclear fuel material can be shortened.

On the other hand, K. FUJIMURA, et. Al., “Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan”, Journal of Power and Energy Systems, Vol.3, No.1, pp.136-145, ( As described in (2009), in light water reactors, particularly when the degree of burnup is high in a reactor loaded with MOX fuel, the ratio of fissionable Pu in the TRU containing MA is as low as about 50%. The ratio is high. If such a TRU is included in the nuclear fuel material in the fuel assembly loaded in the core fuel region, the ratio of MA in the nuclear fuel material contained in this fuel assembly increases. Conversion Engineering Committee Japan Atomic Energy Society, pp. 3-4 to 3-5, as described in February 2004, the absolute value of void reactivity increases. Therefore, measures for reducing the void reactivity, such as shortening the effective fuel length, are required, but this increases costs such as an increase in the core diameter. In addition, there is a method in which the above TRU is used as a nuclear fuel material of FBR by mixing with TRU obtained by reprocessing spent nuclear fuel material containing UO ₂ which is a spent nuclear fuel material of light water reactor. It becomes a cost increase factor.

  An object of the present invention is to provide a fast breeder reactor core capable of reducing the time required for reprocessing spent nuclear fuel material.

An inner core fuel region loaded with a plurality of inner core fuel assemblies, an inner core fuel region, and an outer core fuel region loaded with a plurality of outer core fuel assemblies, an outer core fuel region, and a light water reactor. Spent fuel material in a spent fuel assembly loaded with a core and containing a mixed oxide fuel having a burn-up burnup in the range of 33 GWd / t to 60 GWd / t A blanket region loaded with a blanket fuel assembly having a transuranium element containing plutonium and minor actinides obtained by reprocessing and a nuclear fuel material containing recovered uranium,
The total of the respective proportions of Pu-238 and Pu-240 in the total plutonium of the blanket fuel assembly that stays in the core for 4 cycles and is taken out from the core is 18% by weight or more.

  Obtained by reprocessing spent fuel material in a spent fuel assembly of a fuel assembly containing a mixed oxide fuel that cannot be used for a fuel assembly loaded in the core fuel region within the core of a fast breeder reactor Transuranium elements including plutonium and minor actinides can be effectively used as blanket fuel. In addition, a blanket fuel assembly having a transuranium element containing plutonium and minor actinides obtained by reprocessing the spent fuel material, and a nuclear fuel material containing recovered uranium, and a radial blanket region in the core of the fast breeder reactor. Therefore, when this blanket fuel assembly is taken out from the core as a spent fuel assembly, the nuclear fuel material contained in this spent blanket fuel assembly becomes a nuclear reactor class, and the spent nuclear fuel material is reprocessed. Can be shortened.

  Preferably, the nuclear fuel material contained in the blanket fuel assembly is decoated by thermal load on the cladding of fuel rods contained in the spent fuel fuel assembly, separation of volatile fission products, and reprocessing. Powdered nuclear fuel material produced through each process of pulverizing spent pellet fuel in the fuel rod, and the fuel rod contained in the blanket fuel assembly vibrates the powdered nuclear fuel material into the cladding of the fuel rod It is desirable that the fuel rod be manufactured by filling. Thereby, the possibility of criticality in the reprocessing step is reduced, and the reprocessing step can be simplified.

  Preferably, the nuclear fuel material contained in the blanket fuel assembly is a recycle of spent nuclear fuel material in which the weight ratio of Pu to the heavy metal weight in the spent fuel material in the spent fuel assembly is 3.3% or more. It is desirable to be a nuclear fuel material obtained by processing.

  According to the present invention, since the nuclear fuel material contained in the used blanket fuel assembly becomes a reactor grade, the time required for reprocessing the spent nuclear fuel material can be shortened.

FIG. 2 is a longitudinal sectional view of a half of the core of the fast breeder reactor according to embodiment 1, which is a preferred embodiment of the present invention. FIG. 2 is a cross-sectional view taken at 1/4 of the core of the fast breeder reactor shown in FIG. 1. FIG. 2 is a cross-sectional view of a blanket fuel assembly loaded in a radial blanket region of a core of the fast breeder reactor shown in FIG. 1. It is a characteristic view which shows the relationship between the sum total of each ratio of Pu-238 and Pu-240 shown to all Pu, and the number of stay cycles in the furnace of the fuel assembly loaded to a radial direction blanket area | region. It is a characteristic view showing the relation between the index of the reactor grade Pu based on the composition of Pu contained in the blanket spent nuclear fuel material and the weight ratio of Pu to the heavy metal of the blanket nuclear fuel material.

  Examples of the present invention will be described below.

  The core of the fast breeder reactor according to embodiment 1, which is a preferred embodiment of the present invention, will be described with reference to FIGS. 1, 2 and 3. FIG.

  The core 1 of the fast breeder reactor of the present embodiment has an inner core fuel region 2, an outer core fuel region 3, a radial blanket region 6, and a shield body region 7, which are arranged in order from the core center toward the outer side. The inner core fuel region 2 is disposed at the center of the core 1, and the outer core fuel region 3 surrounds the inner core fuel region 2. The core fuel region of the core 1 includes an inner core fuel region 2 and an outer core fuel region 3. A radial blanket region 6 surrounds the outer core fuel region 3 and a shield region 7 surrounds the radial blanket region 6. The core 1 further has an upper blanket region 4 and a lower blanket region 5 disposed above and below the inner core fuel region 2 and the outer core fuel region 3, respectively. Although not shown, the core 1 is arranged in a reactor vessel of a fast breeder reactor.

  FIG. 2 shows a quarter cross section of the core 1 of the fast breeder reactor. A plurality of inner core fuel assemblies 11 are loaded in the inner core fuel region 2, and a plurality of outer core fuel assemblies 12 are loaded in the outer core fuel region 3. Each of the inner core fuel assembly 11 and the outer core fuel assembly 12 has a plurality of fuel rods, and these fuel rods have a plurality of fuel pellets made of mixed oxide fuel (MOX fuel). Filled. A plurality of blanket fuel assemblies 13 are loaded in the radial blanket region 6. A plurality of shielding bodies 14 are arranged in the shielding body area 7. The cross section of each shielding body 14 is a regular hexagon as in the case of a blanket fuel assembly 13 described later.

  Further, a plurality of control rod assemblies 15 filled with boron carbide are disposed between the inner core fuel assemblies 11 in the inner core fuel region 2. The reactor output of the fast breeder reactor is controlled by taking these control rod assemblies 15 into and out of the core 1.

  As shown in FIG. 3, the blanket fuel assembly 13 includes a plurality of blanket fuel rods 17 and a trumpet pipe 18 that is a stainless steel cylindrical body having a regular hexagonal cross section. The plurality of blanket fuel rods 17 are arranged in an equilateral triangular lattice within the trumpet tube 18. This blanket fuel rod 17 uses a fuel assembly containing a mixed oxide fuel (MOX fuel) taken out from a light water reactor operated with a fuel assembly containing a mixed oxide fuel (MOX fuel) loaded on the core. The inside is filled with nuclear fuel material containing transuranium element (TRU) containing plutonium (Pu) and minor actinide (MA) obtained by reprocessing spent fuel material in the spent fuel assembly, and recovered uranium (U) doing.

The inner core fuel assembly 11 and the outer core fuel assembly 12 are also made of stainless steel having a regular hexagonal cross section, similar to the blanket fuel assembly 13, although the nuclear fuel materials filled in the fuel rods 17 are different. A trumpet tube 18 that is a cylindrical body, and a plurality of fuel rods arranged in an equilateral triangular lattice in the trumpet tube 18 are provided. The fuel rods of the inner core fuel assembly 11 and the outer core fuel assembly 12 were taken out from a light water reactor that was operated by loading the core with a fuel assembly containing uranium oxide (UO ₂ ) fuel as nuclear fuel material. The fuel assembly containing uranium oxide (UO ₂ ) fuel contains Pu or TRU obtained by reprocessing the spent fuel material in the spent fuel assembly, and the nuclear fuel material containing the oxide of deteriorated U. The enrichment degree of fissile Pu (eg, Pu-239) in the outer core fuel assembly 12 is larger than the enrichment degree of fissile Pu (eg, Pu-239) in the inner core fuel assembly 11. .

  Here, the relationship between the sum of the respective ratios of Pu-238 and Pu-240 shown in all Pu and the number of stay cycles in the furnace of the fuel assembly loaded in the radial blanket region will be described with reference to FIG. FIG. 4 shows the number of stay cycles in the reactor of the blanket fuel assembly 13 loaded in the radial blanket region 6 of the core 1 of the fast breeder reactor of the present embodiment, “Proliferation aspects of plutonium recycling”, CR Physique 3, pp B. Pellaud in "1067-1079 (2002)" used as an index for the judgment condition of the reactor grade Pu composition "Pu-240 ratio in total Pu (hereinafter referred to as index 1)" (18% or more is the reactor In consideration of Pu-238 in which the spontaneous fission rate per unit weight and the decay heat are both larger than those of Pu-240, the newly set index, “all It shows the relationship with the sum of the respective ratios of Pu-238 and Pu-240 in Pu (hereinafter referred to as index 2).

  The reason why the new index 2 is used for the Pellaud index 1 in this embodiment will be described below. The greater the spontaneous fission and decay heat of Pu, the lower the reusability of nuclear fuel material containing Pu obtained by reprocessing into weapons. Table 1 shows the spontaneous fission rate and decay heat of Pu isotopes according to the above indices 1 and 2.

The relationship between the spontaneous fission rate and the magnitude of decay heat will be examined using the above indices 1 and 2. For the index 1, 18%, which is the ratio of Pu-240 in all Pu, and for the index 2, 18%, which is the sum of the respective ratios of Pu-238 and Pu-240, in the total Pu. This is a determination condition for a reactor grade. For index 2, the percentage of Pu-238 is X%, and the percentage of Pu-240 is (18-X)%. At this time, spontaneous fission rate SF ₂ is (2.6 × 10 ⁶⁾ on spontaneous fission rate SF ₁ is (0.9 × 10 ⁶⁾ × 0.18, and the index 2 for indicator 1 × X / 100 + (0 . 9 × 10 ⁶ ) × (0.18−X / 100).

Therefore, SF ₂ −SF ₁ = (2.6 × 10 ⁶ ) × X / 100 + (0.9 × 10 ⁶ ) × (0.18−X / 100) − (0.9 × 10 ⁶ ) × 0. 18 = (2.6−0.9 ×) 10 ⁶ × X / 100, and SF ₂ −SF ₁ is larger than 0. Therefore, the spontaneous fission rate is larger when index 2 is used than when index 1 is used, and it can be said that index 2 is a conservative index from the viewpoint of weapon diversion.

Similarly, the decay heat DH ₁ for index 1 becomes 6.8 × 0.18, and the decay heat DH ₂ for index 2 560 × X / 100 + (6.8 ) × (0.18-X / 100) .

Therefore, DH ₂ −DH ₁ = 560 × X / 100 + (6.8) × (0.18−X / 100) −6.8 × 0.18 = (560−6.8) × X / 100 = 553 2 × X / 100, and DH ₂ -DH ₁ is larger than 0. Therefore, the decay heat is larger when index 2 is used than when index 1 is used, and it can be said that index 2 is a conservative index from the viewpoint of weapon diversion.

  From the above explanation, it can be said that the new index 2 is a more conservative index than the index 1 from the viewpoints of both spontaneous fission rate and decay heat.

In the core 1 of the fast breeder reactor of the present embodiment, the inner core fuel assembly 11, the outer core fuel assembly 12, and the blanket fuel assembly 13 are loaded in the core 1 and then stay in the core 1 for 4 cycles. It is taken out from the core 1. In FIG. 4, a characteristic 44 is a spent fuel material in a spent fuel assembly of a fuel assembly containing UO ₂ fuel, which is taken out from a light water reactor operated with a fuel assembly containing UO ₂ fuel loaded in the core. The analysis results for a blanket fuel assembly containing TRU and oxides of deteriorated U as nuclear fuel materials obtained by reprocessing the fuel are shown. Characteristic 45 is operated with a fuel assembly containing MOX fuel loaded in the core. A blanket fuel assembly containing TRU obtained by reprocessing the spent fuel material in the spent fuel assembly including the MOX fuel taken out from the light water reactor and the oxide of deteriorated U as the nuclear fuel material The analysis result is shown. Further, the characteristic 45-1 shown in FIG. 4 is an analysis result when the removal burnup degree of the light water reactor is 33 GWd / t, and a characteristic 45-2 is an analysis result when the removal burnup degree of the light water reactor is 45 GWd / t. 45-3 shows the analysis results when the removal burnup of the light water reactor is 60 GWd / t. Also, the alternate long and short dash line 46 shown in FIG. 4 is a line where the value of index 2 is 18%.

  According to this embodiment, the fuel containing MOX fuel taken out from the light water reactor operated by loading the fuel assembly containing MOX fuel on the core to the blanket fuel assembly 13 loaded in the radial blanket region 6. Since TRU obtained by reprocessing spent fuel material in the spent fuel assembly of the assembly and nuclear fuel material containing oxide of recovered U are used, as is apparent from FIG. Pu contained in the spent nuclear fuel material in the used blanket fuel assembly 13 taken out from the core 1 can be made a nuclear reactor grade. For this reason, the time required for the reprocessing of the spent nuclear fuel material in the used blanket fuel assembly 13 can be shortened. In addition, spent fuel assemblies including MOX fuel that cannot be used as nuclear fuel materials for the inner core fuel assemblies 11 and the outer core fuel assemblies 12 loaded in the core fuel region in the core 1 are used in the spent fuel assemblies. The TRU obtained by reprocessing the fuel material can be effectively used as a blanket fuel.

Furthermore, from a light water reactor operated by loading a fuel assembly containing UO ₂ fuel as a nuclear fuel material into the inner core fuel assembly 11 and the outer core fuel assembly 12 loaded in the core fuel region of the core 1. By using the extracted nuclear fuel material including Pu or TRU obtained by reprocessing the spent fuel material in the spent fuel assembly of the fuel assembly containing UO ₂ fuel and the oxide of deteriorated U, The void reactivity of the core 1 can be reduced to 6 $ or less, and even when a core damage is assumed, the safety of the core 1 can be sufficiently secured. In addition, in the core 1 of this embodiment, a growth ratio of about 1.2 in the middle of the equilibrium cycle can be achieved by loading the UO ₂ fuel, which is mostly oxide of deteriorated uranium, in the radial blanket region 6.

  The core of the fast breeder reactor according to embodiment 2, which is another embodiment of the present invention, will be described below. The core of the fast breeder reactor of the present embodiment has a configuration in which the nuclear fuel material to be filled in each fuel rod 17 of the blanket fuel assembly 13 is powdered instead of pellets in the core 1 of the first embodiment. The other configuration of the core of the fast breeder reactor of the present embodiment is the same as that of the core 1 of the first embodiment.

When reprocessing the spent fuel material in the form of pellets in the spent fuel assembly of the fuel assembly containing the MOX fuel, taken out from the light water reactor operated with the fuel assembly containing the MOX fuel loaded in the core, With this spent nuclear fuel material loaded in the fuel rod, the fuel rod cladding tube is sheared or perforated so that the pelleted spent nuclear fuel material in the fuel rod is easily oxidized. Then, by heating the cladding tube, the pellet-shaped spent nuclear fuel material existing inside is oxidized at a high temperature, and UO ₂ contained in the pellet-shaped spent nuclear fuel material is converted to U ₃ O ₈ . In the process, the fuel pellets in the cladding tube are thermally expanded by heating the cladding tube, and the cladding tube is broken using the tensile stress generated with the increase in the volume of the fuel pellet to perform the decoating. In this embodiment, the coating is performed by applying a heat load to the cladding tube by heating as described above. At the same time, pulverization of the spent nuclear fuel material in the form of pellets and separation of volatile fission products (FP) such as iodine (I) and xenon (Xe), and U ₃ O ₈ containing the remaining FP Is returned to UO ₂ by reduction in a hydrogenated atmosphere and then powdered. The resulting powdered nuclear fuel material is vibrationally filled into a cladding tube to produce several other fuel rods. A blanket fuel assembly 13 is manufactured using these fuel rods, and the plurality of manufactured blanket fuel assemblies 13 are loaded in the radial blanket region of the core of the fast breeder reactor of this embodiment.

  The core of the fast breeder reactor of the present embodiment can obtain each effect produced in the core 1 of the fast breeder reactor of the first embodiment. Further, since the fuel rod used in the blanket fuel assembly 13 loaded in the radial blanket region of the core of the fast breeder reactor of the present embodiment is manufactured by vibrating and filling the powdered nuclear fuel material in the cladding tube, The reprocessing step for obtaining the powdered nuclear fuel material described above is a reprocessing for obtaining the nuclear fuel material used for producing fuel pellets to be filled in the fuel rods of the blanket fuel assembly 13 used in the first embodiment. Compared with the process, the possibility of criticality in the reprocessing process is reduced, and the reprocessing process can be simplified. Incidentally, in the reprocessing step for obtaining the nuclear fuel material used for producing the fuel pellets to be filled in the fuel rods of the blanket fuel assembly 13 used in the first embodiment, the fuel assembly containing the MOX fuel is loaded into the core. It is necessary to extract Pu and TRU by dissolving the spent fuel material in the spent fuel assembly of the fuel assembly containing the MOX fuel taken out from the operated light water reactor with nitric acid. Further, as shown in Table 2, since the solid FP is irradiated with neutrons in the core while being contained in the powdered nuclear fuel material in the blanket fuel assembly 13, technetium 99 (Tc-99) and cesium 135 are received. A long half-life FP such as (Cs-135) can be extinguished in the core.

  The core of the fast breeder reactor according to embodiment 3, which is another embodiment of the present invention, will be described with reference to FIG. The core of the fast breeder reactor according to this embodiment is configured by loading a fuel assembly containing MOX fuel as a nuclear fuel material to be filled in the fuel rods of the blanket fuel assembly 13 loaded in the radial blanket region 6. A nuclear fuel material containing TRU obtained by reprocessing the spent fuel material in the spent fuel assembly of the fuel assembly containing the MOX fuel taken out from the operated light water reactor and the oxide of deteriorated U is used. . In this spent nuclear fuel material, the weight ratio of Pu to the heavy metal weight in the spent nuclear fuel material is 3.3% or more. Other configurations of the fast breeder reactor core of the present embodiment are the same as the configurations of the fast breeder reactor core 1 of the first embodiment.

  The characteristic 54 shown in FIG. 5 shows the reactor grade Pu index 2 based on the composition of Pu contained in the spent nuclear fuel material in the spent blanket fuel assembly 13, and the blanket fuel assembly at the time of loading into the core. It shows the relationship with the weight ratio of Pu to heavy metal in the nuclear fuel material contained in the body. A characteristic 54 is an analysis result.

  Based on the characteristic 54, when the weight ratio of Pu to heavy metal in the nuclear fuel material contained in the blanket fuel assembly is larger, the index 2 of the reactor class Pu is larger, and when the Pu weight ratio is 3.3% or more, It can be seen that the index 2 of the reactor grade Pu is 18% or more, and the conditions for the reactor grade Pu can be satisfied.

  Therefore, the nuclear fuel obtained by reprocessing spent nuclear fuel material in which the weight ratio of Pu to the heavy metal weight in the spent fuel material of the fuel assembly containing MOX fuel is 3.3% or more By loading the blanket fuel assembly 13 containing the material into the radius blanket region 6 to constitute the core of the fast breeder reactor of this embodiment, four cycles have elapsed since the blanket fuel assembly 13 was loaded into the core. Thereafter, the time required for reprocessing the spent nuclear fuel material in the spent blanket fuel assembly 13 taken out from the core can be shortened.

  DESCRIPTION OF SYMBOLS 1 ... Core, 2 ... Inner core fuel area, 3 ... Outer core fuel area, 4 ... Upper blanket fuel area, 5 ... Lower blanket fuel area, 6 ... Radial blanket area, 7 ... Shield body area, 11 ... Inner core fuel Assembly: 12 ... Outer core fuel assembly, 13 ... Blanket fuel assembly, 14 ... Shielding body, 17 ... Blanket fuel rod, 18 ... Trumpet tube.

Claims (3)

An inner core fuel region loaded with a plurality of inner core fuel assemblies; and an inner core fuel region surrounding the inner core fuel region; an outer core fuel region loaded with a plurality of outer core fuel assemblies; and the outer core fuel region; A spent fuel assembly of a fuel assembly loaded in the core of a light water reactor and containing a mixed oxide fuel, wherein the spent fuel assembly is in a range of 33 GWd / t to 60 GWd / t. A radial blanket region loaded with a blanket fuel assembly having a transuranium element containing plutonium and minor actinides obtained by reprocessing the fuel material and a nuclear fuel material containing recovered uranium;
A high speed characterized in that the sum of the respective proportions of Pu-238 and Pu-240 in the total plutonium of the blanket fuel assembly taken out of the core after staying in the core for 4 cycles is 18% by weight or more. Breeder reactor core.   In the reprocessing, the nuclear fuel material contained in the blanket fuel assembly is decoated by a thermal load on the cladding of fuel rods contained in the spent fuel fuel assembly, separation of volatile fission products, and A powdered nuclear fuel material produced through each step of pulverizing spent pellet fuel in the fuel rod, and the fuel rod included in the blanket fuel assembly converts the powdered nuclear fuel material into the fuel rod. The fast breeder reactor core according to claim 1, wherein the core is a fuel rod manufactured by vibration-filling a cladding tube.   The spent nuclear fuel material in which the nuclear fuel material contained in the blanket fuel assembly has a weight ratio of Pu of 3.3% or more with respect to a heavy metal weight in the spent fuel material in the spent fuel assembly. The core of a fast breeder reactor according to claim 1, wherein the nuclear fuel material is obtained by the reprocessing.

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